ORAU Team

NIOSH Dose Reconstruction Project

Technical Information Bulletin: Technical Basis for Estimating the 

Maximum Plausible Dose to Workers at Atomic Weapons 

Employer Facilities

Document Number:

ORAUT-OTIB-0004

Effective Date: 12/04/2003

Revision No.: 02

Controlled Copy No.:_______

Page 1 of 13

Subject Expert: Jeri L. Anderson

Approval: Signature on File Date: 12/04/2003 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 12/04/2003 Richard E. Toohey, Project Director

Approval: Signature on File Date: 12/05/2003

James W. Neton, OCAS Health Science Administrator

Supersedes:

Revision No.: 01

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft Draft 00-A New Technical Information Bulletin to establish a 

process for estimating the maximum plausible 

dose to workers at selected AWE Facilities. 

Initiated by Jeri L. Anderson.

Draft 11/12/2003 00-B Incorporates comments from internal and NIOSH 

review. Initiated by Jeri L. Anderson.

Draft 11/12/2003 00-C Incorporates comments from NIOSH review 

received on redline/strikeout version of TIB dated 

11/12/2003. Initiated by Jeri L. Anderson.

11/20/2003 11/20/2003 00 First approved issue. Initiated by Jeri L. 

Anderson.

Draft 11/24/2003 01-A Incorporates changes to correct errors on pgs. 4 

and 10. TIB incorrectly referred to Absorption 

Type in ingestion discussion. Initiated by Jeri L. 

Anderson.

11/25/2003 11/25/2003 01 Approved issue of revision 01. Initiated by Jeri L. 

Anderson.

Draft 12/02/2003 02-A Adds remainder dose to Table 4. Initiated by Jeri 

L. Anderson.

12/04/2003 12/04/2003 02 Approved issue of revision 02. Initiated by Jeri L. 

Anderson.
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 2 of 13

1.0

The purpose of this document is to provide guidance for estimating the maximum plausible dose to 

workers at Atomic Weapons Employers (AWEs). This document describes an efficiency process that 

may be used to expedite the processing of claims requiring dose reconstruction under the Energy 

Employees Occupational Illness Compensation Program Act (EEOICPA). The exposure matrix in this 

document is designed for estimating the maximum plausible annual dose in all organs with the 

exception of lung, skin, breast, eye, and testes except when the testes dose is used as an analog for 

the prostate. Because the current ICRP model does not calculate a dose to the prostate, the dose to 

the testes is reconstructed and used to determine the probability of causation for prostate cancer. 

This is considered a claimant-favorable approach.

SCOPE

There were approximately 109 AWE facilities that handled only natural uranium in support of the 

atomic weapons program. Of these facilities, 88% operated for 10 years or less and 72% operated for 

5 years or less (see Figure 1). The maximum time operated was twenty-three years and the minimum 

time was less than a year.

Figure 1. Number of AWE facilities versus years of operation.

The processes at these facilities included reduction and recasting, rolling, machining, and extruding of 

uranium, fuel element fabrication, scrap recovery, and recovery of uranium from phosphoric acid. A 

large number of the facilities handled uranium metal. The facilities relevant to this document were 

privately-owned and the AEC work was done on a part-time basis or in addition to their normal 

commercial operations. 

This document applies only to facilities that handled uranium metal and is not to be used for facilities 

that processed thorium, radium, or uranium ores. The intake and dose rates in this exposure matrix 

are to be used in conjunction with the individual worker’s covered employment dates, date of birth, 

and date of cancer diagnosis. If it is necessary to pro-rate intake and dose rates, it is acceptable to 
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 3 of 13

use values for a full year as this is an over-estimation and, thus, claimant-favorable. Because worker 

bioassay and dosimetry data cannot be ignored, the worker’s data, if available, should be reviewed 

and it should be ensured that a dose reconstruction based on this exposure matrix is an overestimate 

and is, therefore, a claimant-favorable approach. 

2.0

During World War II and the time of the Manhattan Engineer District, the permissible level for uranium 

air dust concentration was set at 500 μg/m3 for insoluble uranium compounds and 150 μg/m3 for 

soluble uranium compounds. After the war, the University of Rochester (Rochester, NY) 

recommended lowering the permissible level to 50 μg/m3 for all uranium compounds on the basis of 

chemical toxicity. This level is equivalent to 70 disintegrations per minute per cubic meter (70 

dpm/m3

) and is based primarily on animal studies. The Medical Division of the AEC New York 

Operations Office (NYOO) thought that the “maximum permissible level” should be based on human 

data and was thus unknown. Therefore, the level of 50 μg/m3 was generally referred to as the 

“preferred level” (US AEC 1949). Oftentimes the contractors of the Atomic Energy Commission (AEC) 

used the term “Maximum Allowable (air) Concentration” or “MAC” interchangeably with the “preferred 

level” and reported air-sampling results as multiples of the MAC (NLO 1952, US AEC 1953).

ESTIMATION OF INTERNAL DOSE

In 1949, the Medical Division of the NYOO published a report on the health hazards at seven facilities 

that produced and/or processed uranium for the AEC. These facilities included Mallinckrodt Chemical 

Works, Harshaw Chemical Company, Linde Air Products, Electrometallurgical Company, Simonds 

Saw and Steel Company, Vulcan Crucible Company, and Vitro Manufacturing Company. These 

facilities were the earliest of those constructed by the Manhattan Engineer District and, with the 

exception of Vulcan Crucible, were very large operations in comparison to the facilities for which this 

document is relevant. Up until the time of this report, surveys by the NYOO indicated that out of 648 

exposed workers at these plants, 9% were exposed to uranium air concentrations greater than 125 

MAC (> 6250 μg/m3

), 9% were exposed at 25-125 MAC (1250-6250 μg/m3), and 82% were exposed 

to less than 25 MAC (< 1250 μg/m3

). As a result of this report, significant improvements were made in 

the operational conditions such as re-design of ventilation systems, enclosing some operations, and 

using remote control (US AEC 1949). By the end of 1949, exposure levels were significantly reduced 

at these larger plants even though production levels increased (Mason 1958).

Vulcan Crucible (Aliquippa Forge) was a steel mill that contracted with the AEC to roll uranium billets 

into rods on a part-time basis from 1948 to 1949. The contract (AEC Contract No. AT-(30-1)-407) 

stated that the plant was to arrange to spend “at least two consecutive weeks out of every five 

consecutive weeks” performing the AEC contract work. Four of the most exposed workers in the 

1949 study were from this plant and the report stated that discussions were under way to shut down 

this operation (US AEC 1949). Operations were shut down in late 1949 and the plant was 

decontaminated to 1950 standards (Adams and Payne 1992).

It is important to note that these early exposure studies were not very selective in terms of particle 

size (Stannard 1988). Also, the air concentration measurements represented the amount uranium in 

the air where the workers were located but not necessarily what was taken into the lung or the body.

For the purpose of estimating maximum plausible dose, it is assumed that a worker receives a 

constant exposure to a maximum level of 5000 ug/m3 (100 MAC) during the entire period of operation. 

This means that it is assumed that the worker is exposed at this level for eight hours per day, five 

days per week, 50 weeks per year, whereas time-weighted average studies of even the larger plants 

like Mallinckrodt show that the majority of time workers were exposed at lower levels this. It is 

assumed the worker is exposed at this level during lunches, breaks, and when working at tasks that 
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 4 of 13

are not AEC-related. This is considered claimant-favorable because the operations at the facilities for 

which this document is applicable were smaller, mostly part-time operations. This maximum exposure 

level is converted to pCi/m3 by multiplying by 0.677 pCi/μg, which is the specific activity of natural 

uranium.

The annual amount inhaled due to chronic exposure is determined by multiplying the maximum air 

concentration of 3400 pCi/m3 (5000 μg/m3

) by the ICRP 66 recommended breathing rate of 1.2 m3/h 

and by an assumed 2000-hour work year. This results in an annual intake of 8.1E+06 pCi (1.2E+07 

μg). This value for the cumulative annual intake is assumed to result from a chronic exposure and is 

used to determine the annual internal organ dose by applying the current ICRP models. The 

Absorption Type is assigned to Type M as recommended by the ICRP in Publication 71 (ICRP 1995b) 

for uranium compounds for which there is no specific information on solubility and absorption. 

Although human and animal studies have indicated that uranium oxides can be relatively insoluble, 

the selection of Type M results in greater organ uptake and is thus considered claimant favorable. 

ICRP Publication 66 default values (ICRP 1994) should be selected for particle deposition 

parameters. The annual organ doses are entered into the NIOSH IREP program assuming a chronic 

exposure and a constant distribution. The dose is attributed to alpha radiation.

Internal exposure can also be caused by individuals ingesting material by smoking or eating while 

their hands are contaminated with uranium. To estimate the quantity of this ingested material, several 

claimant favorable assumptions have been made. 

1. The individual’s hands have become contaminated from contact with the most highly 

contaminated surfaces in the facility, the floor. The individual’s hand is assumed to become 

contaminated at a level equal to 10% of the level of contamination on the floor.

2. A study was done at the Oak Ridge Gaseous Diffusion Plant (ORGDP) to determine the intake 

of uranium from hand contamination. The study indicated that the amount of uranium that is 

transferred from the hand to the cigarette while smoking was approximately 1% of the material 

on the surface of the hand (Bailey 1958). For this estimate, 10% was used as a bounding 

estimate to include all forms of consumption.

3. The individuals hand is to be approximated by an area of 4” by 6” and is evenly contaminated.

When the estimated surface contamination level of 8.1E+7 pCi/m2 (see section 3.0 below) is 

multiplied by the surface area of the hand (0.0155 m2

) and the 10% transfer factor is applied, the total 

activity on the individual’s hand is equal to 1.26E+5 pCi. If 10% of this is ingested on a daily basis, 5 

days per week for 50 weeks, the total annual ingestion of uranium can be calculated to be 3.14E+6 

pCi annually. 

The cumulative annual ingestion intake is assumed to result from a chronic exposure for the entire 

time the individual was employed at the facility during the uranium contract period and is used to 

calculate the annual internal organ doses by applying the current ICRP models. The f1 value is 

assigned to 0.02 in Publication 69 (ICRP 1995a) for uranium compounds for which there is no specific 

information on solubility. The annual organ doses are entered into the NIOSH IREP program 

assuming a chronic exposure and a constant distribution. The dose is attributed to alpha radiation.

If eating or drinking occurred in the production portions of a facility, airborne radioactivity can settle 

out on food or drink. This causes an ingestion mode that is not specifically associated with 

transferring the material from the hand. A realistic estimate of this would have to involve the settling 

rate of the airborne contamination onto the food or drink. The surface contamination estimate (see 
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 5 of 13

Table 1. Most probable dimensions of the 

uranium ingots used in processing (Bonfer 

2003).

Shape Dimensions (inches)

Rectangular 24 L × 16 W × 4 H

Cylindrical 20 L × 13 Dia.

section 3.0) used this settling rate and accumulated material over the course of a year with no 

removal mechanisms factored in. Since the surface contamination level was the starting point for the 

ingestion estimate, it inherently accounts for an entire years worth of settling. Assuming food and 

drink are consumed at some point during the day that they are brought into the area, the amount of 

material ingested would be a fraction of one days worth of settling. Since a fraction of one years 

worth of settling is already accounted for, it is believed that this ingestion estimate accounts for 

ingestion from all modes.

3.0

The maximum size of the uranium ingots used in the atomic weapons program was used to estimate 

external exposure during processing. The dimensions of the two larger ingots are shown below in 

Table 1.

ESTIMATION OF EXTERNAL DOSE

It was assumed that the most exposed workers were 

the machinists and general laborers so the exposure 

geometry was assumed to be 75% anterior-posterior 

(AP) and 25% isotropic (ISO) as recommended in the 

NIOSH External Dose Reconstruction 

Implementation Guideline (US DHHS 2002a). The 

workers were assumed to spend 7 hours per day 

performing their duties and 1 hour per day at lunch and on breaks. Although during performance of 

the job, the worker could have been any distance from the source (the ingots), to be claimantfavorable, it was assumed the worker spent the entire 7 hours 1 foot from the source. Also, to be 

claimant-favorable, it was assumed that the worker spent his/her lunch and breaks in close proximity 

to an ingot.

For the calculations of exposure rate, it was assumed that the ingots were composed entirely of 

natural uranium metal that had been allowed to decay for 100 days resulting in significant dose 

contribution from the Bremsstrahlung radiations created by the Pa-234m and Th-234 beta emissions. 

A large percentage of the exposure is from Bremsstrahlung photons with energy between 0.8 and 1 

MeV. The exposure rate at various distances from the ingots was calculated using MicroShieldTM 

(Grove Engineering 2003) and MCNP (LANL 2003). The calculated values are shown in Table 2.

Table 2. Calculated exposure rates from uranium metal ingots.

Shape of ingot

Distance 

from surface

MicroShieldTM

exposure rate (mR/hr)

MCNP dose rate 

(mrem/hr)

Rectangular 1 cm 7.90 8.18

1 ft 1.91 2.04

1 m 0.29 0.36

Cylindrical 1 cm 8.27 8.30

1 ft 1.13 1.14

1 m 0.17 0.18

The MCNP calculated dose rates were higher than the MicroShield dose rates for both the rectangular 

and the cylindrical ingot. It was assumed that the majority of the worker’s time was spent at 1 foot 

from the surface of the ingot. The calculated dose rates at 1 foot and 1 meter from the rectangular 

ingot were the highest so those values were used in the dose estimation. The MCNP-calculated dose 

rate was multiplied by the number of hours spent in the appropriate exposure geometry, and then 

multiplied by 250 days per year (8 hours per day for a total of 2000 hours per year). The organ doses 

were calculated by multiplying the calculated annual dose rate by the “Ambient Dose Equivalent 
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 6 of 13

(H*(10)) to Organ Dose (HT)” photon dose conversion factors found in Appendix B of the NIOSH 

External Dose Reconstruction Implementation Guideline (DHHS 2002a). The annual dose rate was 

divided evenly between the photons with energy between 30 and 250 keV and photons with energy 

greater than 250 keV. Table 3 shows the calculated annual organ doses due to exposure to a natural 

uranium ingot.

Table 3. Annual organ doses due to external exposure to a 

natural uranium ingot.

Organ

Annual organ dose (rem)

Photons with

E=30-250 keV

Photons with

E>250 keV Total

Urinary bladder 1.9 2.3 4.2

Bone (red marrow) 1.6 2.2 3.8

Bone (surfaces) 2.6 2.2 4.8

Colon 1.8 2.3 4.1

Esophagus 1.5 2.2 3.7

Gonads (ovaries) 1.8 2.2 4.1

Gonads (testes) 2.1 2.4 4.4

Liver 1.9 2.3 4.2

Remainder 1.8 2.2 4.0

Stomach 1.9 2.3 4.2

Thymus 1.9 2.2 4.1

Thyroid 2.0 2.4 4.4

Uterus 1.8 2.1 3.9

The doses in Table 3 are entered into the NIOSH IREP program assuming a chronic exposure and a 

constant distribution. The doses in the second column of Table 3 are attributed to photons with 

energy between 30 and 250 keV and the doses in the third column are attributed to photons with 

energy greater than 250 keV.

It was also assumed that workers were exposed to contaminated surfaces during their entire work 

period. The level of surface contamination was determined by first calculating a terminal settling 

velocity for 5-μm activity mean aerodynamic diameter (AMAD) particles. The calculated terminal 

settling velocity of 0.00075 meters per second is multiplied by the assumed maximum air 

concentration of 3400 pCi/m3

. It was assumed that due to the constant air concentration, the surface 

contamination level was due to 365 days of constant deposition. This is considered claimantfavorable as facility housekeeping practices most likely kept levels lower than this. At the assumed air 

concentration level with the calculated terminal settling velocity, a 365-day buildup will result in a 

contamination level of 8.1E+07 pCi/m2

. The annual organ dose due to exposure to contaminated 

surfaces was determined by multiplying the surface contamination level by the dose coefficients for 

contaminated ground surfaces for U-238 and daughters Pa-234m and Th-234 from Federal Guidance 

Report No. 12 (US EPA 1993). These dose conversion factors were modeled using an 

anthropomorphic phantom standing on an isotropic plane source in air. Table 4 shows the annual 

external organ dose estimates.

The doses in Table 4 are entered into the NIOSH IREP program assuming a chronic exposure and a 

constant distribution. The dose is divided equally between photons with energy between 30 and 250 

keV and photons with energy greater than 250 keV.

The assumption was also made that workers received an annual occupationally related diagnostic xray. The exposure geometry was assumed to be posterior-anterior (PA) (DHHS 2002). The air kerma 

at skin entrance for the diagnostic chest x-ray was estimated to be 0.108 R (Scalsky 2003). The 
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 7 of 13

Table 4. Annual organ doses due to external 

exposure to ground surfaces contaminated with 

natural uranium dust.

Organ Annual dose (rem)

Adrenals 4.0E-02

Urinary bladder 4.4E-02

Bone surface 1.1E-01

Brain 4.2E-02

Esophagus 3.7E-02

Stomach wall 4.4E-02

Small intestine wall 4.1E-02

Upper large intestine wall 4.2E-02

Lower large intestine wall 4.3E-02

Kidney 4.4E-02

Liver 4.4E-02

Muscle 5.4E-02

Ovaries 4.1E-02

Pancreas 3.9E-02

Red Marrow 4.5E-02

Remainder 4.3E-02

Spleen 4.4E-02

Testes 5.6E-02

Thymus 4.3E-02

Thyroid 4.8E-02

Uterus 4.1E-02

organ doses were calculated by multiplying the air kerma by the “Kerma (Ka) to Organ Dose (HT)” 

photon dose conversion factors for 30-250 keV photons found in Appendix B of the NIOSH External 

Dose Reconstruction Implementation Guideline (DHHS 2002). Table 5 below shows the annual organ 

doses due to the assumed annual diagnostic chest x-ray. The values in Table 5 are entered into the 

NIOSH-IREP program as the annual dose due to an acute exposure to photons (E=30-250 keV) and 

a constant distribution is assumed. 

Table 5. Annual organ doses due to the 

assumed annual diagnostic chest x-ray.

Organ Annual dose (rem)

Bladder 0.074

Red bone marrow 0.124

Bone Surface 0.178

Breast 0.059

Colon 0.095

Esophagus 0.105

Eye 0.022

Ovaries 0.110

Testes 0.061

Liver 0.101

Lung 0.133

Remainder organs 0.110

Skin 0.110

Stomach 0.077

Thymus 0.048

Thyroid 0.052

Uterus 0.096
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 8 of 13

4.0

During operations with uranium, airborne contamination can be caused directly from operations with 

uranium bearing materials as well as by the resuspension of surface contamination. This airborne 

contamination can be removed by settling on floors or other surfaces or by ventilation (both 

engineered and natural building ventilation). The surface contamination levels can be increased by 

the airborne contamination settling or directly by operations with uranium. This level is decreased by 

various incidental means such as general housekeeping or tracking of contamination out of the area 

on equipment or personnel.

ESTIMATION OF EXPOSURE TO RESIDUAL RADIOACTIVITY

For the purpose of estimating the airborne contamination from residual contamination, it is assumed 

that during uranium operations, these factors reached an equilibrium that caused the airborne 

concentration and surface contamination to remain at a constant level. Once the generation source 

from uranium operation was removed, the other factors then caused the over mass of uranium to 

decrease at some rate due to the remaining removal factors.

This analysis assumes that 1) uranium operations directly accounted for half the airborne 

concentration in the facility during operations and 2) all the remaining removal mechanisms eliminated 

only 1% of the uranium mass per day. With these assumptions in place, an equation can be written 

for the airborne concentration as a function of time.

t tA MAC e *01.0 )( 50 * − =

Where:

A(t) = Airborne activity at time t (pCi/m3

) 

t = time since uranium operations ended (days)

MAC = Maximum Allowed Concentration (34 pCi/m3

) 

When this is multiplied by the breathing rate of 9.6 m3 per day, the daily intake is calculated. This 

equation can then be integrated from the end of uranium operations to infinity to determine the total 

amount of uranium that could be inhaled from residual contamination. This integration results in:

I = *6.9 50MAC .0/ 01

Where:

I = total intake from inhalation of residual activity (pCi)

MAC = Maximum allowed concentration (34 pCi/m3) 

9.6 = daily breathing rate (m3/day)

0.01 = fraction of material removed per day

The results of this equation indicate a total of 1.63x106 pCi would be inhaled. This equates to 

approximately 20% of the 8.1x106 pCi annual intake during operations. Therefore, the total inhalation 

intake from residual contamination can be overestimated by assuming one additional year of 

inhalation exposure at the operational level of 100 MAC. 

The external radiation caused by residual contamination can be overestimated by assuming the 

contamination levels that existed during operations continued to exist after operations ceased. The 

annual dose from this source of radiation has already been estimated and can be found in Table 4. 
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 9 of 13

This dose should be added to an individuals estimate for each year the individual worked at the facility 

following the end of uranium operations.

Ingestion of material is most closely related to removable surface contamination rather than fixed 

surface contamination. Airborne concentrations are also related to removable surface contamination. 

Therefore, the removable contamination levels are assumed to decrease as the same rate as the 

airborne concentrations. For this reason, the same approach will be taken for ingestion estimates 

following periods of operations. That is, the ingestion intakes from residual contamination will be 

estimated by assuming one additional year of ingestion at the same rate it occurred during operation.

5.0

Adams, W.C., and Payne, J.L., “Characterization Survey of Portions of the Aliquippa Forge Site, West 

Aliquippa, Pennsylvania,” prepared for the Office of Environmental Restoration, U.S. 

Department of Energy, ORISE 93/A-6, December 1992 

REFERENCES

Bailey, J.C., “Personnel Contamination as a Uranium Hazard,” in Symposium on Occupational Health 

Experience and Practices in the Uranium Industry, HASL-58, Health and Safety Laboratory, 

U.S. Atomic Energy Commission, New York Operations Office, October 15-17, 1958

Bonfer, D., Personal Communication, 2003

Grove Engineering, MicroShield Software Program, Version 6.01, Copyright 1992-2003

Heatherton, R.C., and Quigley, J.A., AEC Radiation Exposure Report-1959, February 11, 1960 [found 

on pg. 34 of O:\DOE Site Images\FEMP\09-17-03 Data Capture\010001640.pdf]

International Commission on Radiological Protection (ICRP), “Human Respiratory Tract Model for 

Radiological Protection,” ICRP Publication 66, pg. 198, 1994

International Commission on Radiological Protection (ICRP), “Age-dependent Doses to Members of 

the Public from Intake of Radionuclides: Part 3 Ingestion Dose Coefficients, ICRP Publication 

69, pg. 65, 1995a

International Commission on Radiological Protection (ICRP), “Age-dependent Doses to Members of 

the Public from Intake of Radionuclides: Part 4 Inhalation Dose Coefficients, ICRP Publication 

71, pg. 299, 1995b

LANL, X-5 Monte Carlo Team, “MCNP—A General Monte Carlo N-Particle Transport Code, Version 

5,” LA-UR-03-1987, Los Alamos National Laboratory, April 24, 2003

Mason, M.G., “A Summary of Fifteen Years of Experience With Dust Problems in Refining and 

Fabrication of Uranium,” in proceedings of the Symposium on Occupational Health Experience 

and Practices in the Uranium Industry, HASL-58, October 15-17, 1958

National Council on Radiation Protection and Measurements (NCRP), Permissible Dose from External 

Sources of Ionizing Radiation, Including Maximum Permissible Exposure to Man, NCRP 

Report No. 17, National Bureau of Standards Handbook 59, 1954
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 10 of 13

National Lead Company of Ohio (NLO), Production Report on the Rolling of Two Hundred and 

Eighteen Uranium Billets at Bethlehem Steel Corporation’s Lackawanna Plant on March 15, 

1952, Contract No. AT(30-1)-1156, March 21, 1952

Scalsky, E.D., Technical Basis Document for the Savannah River Site To be Used for EEOICPA Dose 

Reconstructions, ORAUT-TKBS-003, Effective Date August 21, 2003

Stannard, J.N., Radioactivity and Health: A History, Volume 1, Laboratory Research

US AEC, “Health Hazards in NYOO Facilities Producing and Processing Uranium: A Status ReportApril 1, 1949,” Prepared by the New York Operations Office Medical Division, Issued April 18, 

1949

, prepared for the 

U.S. Department of Energy, Office of Health and Environmental Research, DOE/RL/01830-

T59, October 1988

US AEC, “Simonds Saw and Steel Company, Occupational Exposure to Uranium,” New York 

Operations Office Health and Safety Division, February 18, 1953

US DHHS, External Dose Reconstruction Implementation Guideline, Rev. 1, National Institute for 

Occupational Safety and Health, August 2002 (2002a)

US DHHS, Progress Report on Residual Radioactive and Beryllium Contamination at Atomic 

Weapons Employer Facilities and Beryllium Vendor Facilities, Prepared by the National 

Institute for Occupational Safety and Health, November 2002 (2002b)

US EPA, External Exposure to Radionuclides in Air, Water, and Soil; Federal Guidance Report No. 

12, September 1993
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 11 of 13

Appendix

List of Applicable Facilities

Site name Years of operation

Ajax Magnethermic Corp. 1958–1962

Alba Craft 1952–1957

Aliquippa Forge 1947–1950

Allegheny-Ludlum Steel 1950–1952

Allied Chemical and Dye Corp. 1951–1969

Allis-Chalmers Co. 1943–1944

Aluminum Co. of America (NJ) 1944

Aluminum Co. of America (PA) 1944–1945

AMCOT 1961–1962

American Bearing Corp. 1954–1959

American Chain and Cable Co. 1944

American Peddinghaus Corp. 1978

American Potash & Chemical 1959–1961

Anaconda Co. 1942, 1956–1959

Armco-Rustless Iron & Steel 1948

Armour Fertilizer Works 1951–1955

Associated Aircraft Tool and Manufacturing 1956

B&T Metals 1943

Baker and Williams Co. 1957–1962

Baker and Williams Warehouses 1940–1950

Baker Brothers 1943–1944

Baker-Perkins Co. 1956

Bell Telephone Laboratories 1943–1944

Bendix Aviation (Pioneer Division) 1960

Besley-Wells 1953

Bethlehem Steel 1949–1952

Birdsboro Steel & Foundry 1951–1952, 1962

Bliss & Laughlin Steel 1948–1952

Blockson Chemical Co. (Bldg. 55) 1952–1962

Bloomfield Tool Co. 1947, 1951

Bridgeport Brass Co., Havens Laboratory 1952–1962

Brush Beryllium Co. (Detroit) 1940–1955

BWX Technologies (Virginia) 1959, 1969–1972, 1985–2001

C.H. Schnoor 1943–1951

C.I. Hayes, Inc. 1964

Callite Tungsten Co. 1944

Carboloy Co. 1956

Carpenter Steel Co. 1943–1944

Chambersburg Engineering Co. 1957

Chapman Valve 1948–1949

Cincinnati Milling Machine Co. 1963

Combustion Engineering 1965–1972

Copperweld Steel 1943–1946

Dow Chemical Co. (Madison Site) 1957–1960

Electro Circuits, Inc. 1952–1953

Fenn Machinery Co. 1950
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 12 of 13

List of Applicable Facilities (Continued)

Site name Years of operation

Fenwal, Inc. 1967–1968

Frankford Arsenal 1952–1954

Gardinier Inc. 1951–1954, 1956–1961

Granite City Steel 19581966

Gruen Watch 1956

Heald Machine Co. 1960

Heppenstall Co. 1955

Herring-Hall Marvin Safe Co. 1943–1951

Hunter Douglas Aluminum Corp. 1959–1963

Huntington Pilot Plant 1951–1963

International Minerals and Chemical Corp. 1951–1961

International Nickel Co., Bayonne Laboratories 1951–1952

International Register 1943

Ithaca Gun Co. 1961–1962

J.T. Baker Chemical Co. 1948, 1957–1958

Jessop Steel Co. 1950–1954

Joslyn Manufacturing and Supply Co. 1944–1952

Kaiser Aluminum Corp. 1959

Kerr-McGee 1962–1973

La Pointe Machine and Tool Co. 1956

Landis Machine Tool Co. 1952

Magnus Brass Co. 1954–1957

Mathieson Chemical Co. 1951–1953

Medart Co. 1951–1952

Metals and Controls Corp. 1952–1967

Mitchell Steel Co. 1954

National Research Corp. 1944–1952

New England Lime Co. 1963

New York University 1946–1952

Oliver Corp. 1956–1957, 1961–1962

Picatinny Arsenal 1947–1948

Podbeliniac Corp. 1957

Precision Extrusion Co. 1949–1950, 1956–1959

Quality Hardware and Machine Co. 1944–1945

R.W. Leblond Machine Tool Co. 1961

Revere Copper and Brass 1943–1955

Roger Iron Co. 1956

Sciaky Brothers, Inc. 1953

Seymour Specialty Wire 1962–1964

Southern Research Institute 1955–1958, 1962, 1976

Spencer Chemical Co., Jayhawks Works 1958–1963

Sperry Products, Inc. 1952–1953

Star Cutter Corp. 1956

Stauffer Metals, Inc. 1961

Superior Steel Co. 1952–1957

Tennessee Valley Authority 1951–1955

Texas City Chemicals 1952–1956

Titus Metals 1956
Effective Date: 12/04/2003 Revision No. 02 Document No. ORAUT-OTIB-0004 Page 13 of 13

List of Applicable Facilities (Continued)

Site name Years of operation

Tocco Induction Heating Div. 1968–1969

Torrington Co. 1951–1953

Tube Reducing Co. 1952, 1957

U.S. Steel Co., National Tube Division 1959–1960

United Nuclear Corp. 1958–1969

University of Michigan 1944

Ventron Corporation 1942–1948

Virginia-Carolina Chemical Corp 1952–1957

Vulcan Tool Co. 1959

W.E. Pratt Manufacturing Co. 1943–1946

W.R. Grace (Tennessee) 1958–1970

W.R. Grace Co., Agricultural Chemical Div. 1954

Westinghouse Atomic Power Development Plant 1941–1944

Wyckoff Drawn Steel Co. 1943

Wyckoff Steel Co. 1950